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All sourcesNASA NTRSDOE OSTI
NASA NTRS1993

Preliminary Stratigraphic Basis for Geologic Mapping of Venus

The age relations between geologic formations have been studied at 36 1000x1000 km areas centered at the dark paraboloid craters. The geologic setting in all these sites could be characterized using only 16 types of features and terrains (units). These units form a basic stratigraphic sequence (from older to younger: (1) Tessera (Tt); (2-3) Densely fractured terrains associated with coronae (COdf) and in the form of remnants among plains (Pdf); (4) Fractured and ridged plains (Pfr); (5) Plains with wrinkle ridges (Pwr); (6-7) Smooth and lobate plains (Ps/Pl); and (8) Rift-associated fractures (Fra). The stratigraphic position of the other units is determined by their relation with the units of the basic sequence: (9) Ridge bells (RB), contemporary with Pfr; (10-11) Ridges of coronae and arachnoids annuli (COar/Aar), contemporary with wrinkle ridges of Pwr; (12) Fractures of coronae annuli (COaf) disrupt Pwr and Ps/Pl; (13) Fractures (F) disrupt Pwr or younger units; (14) Craters with associated dark paraboloids (Cdp), which are on top of all volcanic and tectonic units except the youngest episodes of rift-associated fracturing and volcanism; (15-16) Surficial streaks (Ss) and surficial patches (Sp) are approximately contemporary with Cdp. These units may be used as a tentative basis for the geologic mapping of Venus including VMAP. This mapping should test the stratigraphy and answer the question of whether this stratigraphic sequence corresponds to geologic events which were generally synchronous all around the planet or whether the sequence is simply a typical sequence of events which occurred in different places at diffferent times.

Basilevsky, A. T.↗
DOE OSTI2016

RELAP-7 Pressurizer Component Development Updates

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). As part of the efforts to expand the capability for PWR simulation, an equilibrium single-region pressurizer model has been implemented in RELAP-7. The pressurizer component can simulate pressure and water level change through insurge, spray, and heating processes. Two simple tests – one for insurge process and another for outsurge process – have been reported to demonstrate and verify the functions of the pressurizer model. The typical single-phase PWR system model presented in the first RELAP-7 milestone report has been updated, as part of system level test for the new pressurizer model. The updated PWR system model with the pressurizer component can be used for more realistic transient simulations. The addition of the equilibrium single-region pressurizer model represents the first step of developing a suite of pressurizer models with different levels of fidelities. Non-equilibrium two-region two-zone model and non-equilibrium five-region three zone model will be implemented in the near future. Cross comparison and validation will be pursued when higher fidelity pressurizer models are completed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Applicable Experiment Selection

The general method for performing validation of as loaded criticality safety calculations using UNF ST&DARDS is presented in a paper by Clarity, which includes a description of the UNF-ST&DARDS system. Proof-of-principle analyses were performed in the summer of 2019 for MPC-32 dual purpose canisters (DPCs) containing pressurized water reactor (PWR) fuel assemblies. Summaries of these results are presented in this and a companion paper for this conference. The current paper describes the TSUNAMI-IP calculations performed to select applicable experiments for validation of 11 MPC-32 DPCs. The companion paper discusses the TSUNAMI-3D calculations used to generate sensitivity data to support the experiment selections discussed here. Experiment selection is based on the sensitivity/uncertainty (S/U) methods used to validate criticality safety calculations of as-loaded DPCs containing pressurized water reactor (PWR) spent nuclear fuel (SNF). This process has been demonstrated and is summarized in this paper. The approach is similar to that used in NUREG/CR-7109, which provides an approach for validation of PWR burnup credit (BUC), including major and minor actinides and major fission products. The premise of S/U-based validation is that applicable experiments—those having a similar bias to a given application system—will have similar sensitivities for each isotope and reaction in the two systems. It is assumed that cross sections with larger uncertainties are more likely to contain data errors which contribute to the bias. The integral index c k thus propagates the system sensitivities with the nuclear covariance data to calculate a correlation coefficient representing the similarity of the two systems. In this work, a c k value of 0.8 or higher is interpreted as identifying an experiment with sufficient similarity for use in validation. This paper presents a brief summary of the characteristics of the 11 MPC-32 DPCs used in the proof of-principle analysis for as-loaded criticality safety calculation validation and an overview of the critical experiment suite with which each of these DPC models was compared. A summary and discussion of c k results is also presented, followed by conclusions and a discussion of future work to be performed for validation of UNF ST&DARDS as-loaded criticality safety calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Understanding of corrosion mechanisms after irradiation: effect of ion irradiation of the oxide layers on the corrosion rate of M5 - 2016-0081

Irradiation damage in fuel cladding material is mainly caused by the neutron flux that results from fission reactions occurring in the fuel. To avoid the constraints inherent in handling radioactive material, the irradiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. We performed an original experiment using ion irradiation to more specifically study the influence of irradiation damage in the oxide on the corrosion rate of M5{sup R}. It has been established that irradiation with a 1.3-MeV helium ion at a fluence of 10{sup 17} cm{sup -2} results in significant modifications of oxide properties, oxygen diffusion flux, and oxidation kinetics, as evidenced by Raman spectroscopy, secondary ion mass spectrometry (SIMS) analyses, and measurements of mass gains. A newly identified Raman vibration band at 712 cm{sup -1} was linked to the presence of irradiation defects and allowed the evolution of their concentrations to be followed. The oxygen diffusion flux was significantly reduced after irradiation partly due to a surface concentration decrease of oxygen. The defects remained present in the oxide after 100 days of annealing in pressurized water reactor (PWR) conditions and were thus very stable in PWR conditions, which probably means that these defects would be stable in the reactor. According to the kinetics and in agreement with the results obtained by SIMS analyses, the oxidation rate was significantly reduced after ion irradiation, and this effect remained beyond 100 days in agreement with the high stability of irradiation defects in PWR conditions. An original model described quite well the oxidation kinetic results. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2018

Implementation of ANL’s Mechanics Based Evolutionary Fatigue Modeling Through ABAQUS-WARP3D Based High-Performance Computing Framework

This report presents an update on the environmental fatigue research that is being conducted at Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. Argonne is developing a fully mechanistic fatigue evaluation approach without using empirical fatigue (S~N) curves. This approach is based on the fundamental concept of the time evolution of progressive fatigue damage rather than on the conventional S~N curve approaches using end-of-life data. In FY 2017, we performed extensive validation of this approach with respect to fatigue test data for 316 stainless steel [1]. This validation was performed for different loading cases, including constant, variable, and random amplitude. In the present FY 2018 semi-annual report, we present the further advances of Argonne’s environmental fatigue research work in the context for more practical applications. In this report, we discuss a methodology for fully mechanistic (i.e., not using S~N curves) fatigue life evaluation of reactor components subjected to realistic loading cycles, namely, design-basis loading cycles. The loading cycles include plant heat-up, full-power, and cool-down operations. As a test case, we considered a typical pressurized water reactor surge line, which is made of 316 SS. To perform the fatigue simulation for thousands of fatigue cycles in a computationally cost effective way, we modified our previous desktop-based finite element (FE) modeling approach to work in a high-performance computing (HPC) framework. For the HPC implementation, we developed a hybrid framework based on commercial FE software (ABAQUS), open-source FE software (WARP3D), and Argonne-developed evolutionary cyclic-plasticity modeling methods. We validated this HPC-based cycle-by-cycle damage model for the entire fatigue life of a Pressurized Water Reactor (PWR) surge line (SL) pipe with respect to assumed loading cycles. The simulated fatigue life was found to be 5855 cycles, which is 85% accurate as compared to the corresponding small-specimen-based experimental fatigue life (6914 cycles). Also, the simulated stress history captures the cyclic hardening and softening behavior of the material for entire fatigue cycles. The FE simulation of the PWR SL pipe was conducted in a reasonable time of 12.5 days. These results show the promise that a fully mechanistic (not using S~N curves) fatigue life evaluation of a safety-critical nuclear reactor component (or even other safety critical components like those in aircraft, aero-engines, etc.) is possible. We anticipate that this type of methodology will drastically reduce the uncertainly associated with conventional fatigue life estimates based on empirical S~N curves. We also proposed an FE model that is based on a hybrid full-component and single-element approach and that can readily be used by industry if HPC resources are not available. In this approach, a single-cycle FE simulation has to be performed first for the required loading cycle. Then, the resulting strain/stress profile at the hotspot (highest strain/stress location) is used as input strain/stress loading to the single-element FE model. The single-element FE model is run under multiple fatigue cycles as long as the failure criteria are achieved. The cycle number at failure gives the fatigue life of the component in question. This cost-effective framework does not require expensive HPC hardware and can be performed in a reasonable time by using a desktop computer. We demonstrated this approach with respect to PWR SL fatigue evaluation under two realistic design-basis loading cycles (simple and detailed). The associated results were validated with respect to test data obtained through fatigue testing of uniaxial specimens under the in-air condition at 300 °C. The approach is in the development stage and requires further improvement and validation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Generic Pressurized Water Reactor Model for SAPHIRE

Generic Pressurized Water Reactor Model for SAPHIRE This report contains a data dump from a SAPHIRE Generic Pressurized Water Reactor (PWR) model. The intent is to have this information reviewed for export control and classification so that the generic model can be used external to INL. The facility model is not applicable to any existing nuclear reactors. The Generic PWR model does not represent any existing power plant and was created to support research and training purposes. Three distinct types of data are used to create a SAPHIRE risk model: - Event trees representing typical PWR sequences - Fault trees representing generic system performance - Basic events (e.g., component failure information) Where possible, we have taken information from publicly available documents and referenced that information tagged to a specific item.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

Experimental validation of the DARWIN2.3 package for fuel cycle applications

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes nuclear data for the next library JEFF3.2. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Neutronics studies of uranium-based fully ceramic micro-encapsulated fuel for PWRs

This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The benefits of a fast reactor closed fuel cycle in the UK

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

CESAR5.3: An Industrial Tool for Nuclear Fuel and Waste Characterization with Associated Qualification - 12067

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of cooling time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR packages, while 12-PWR packages could be emplaced after 40 years or less. HLW from reprocessing LWR UOX fuel or metal fuel from fast reactors, could be disposed of in salt after 10 to 50 years of decay storage depending on the specific composition and other factors. For the deep borehole disposal concept no near-field temperature limits are recognized because no performance credit is taken for waste form or waste package integrity, or containment by the near-field host rock. These results show the key differences in thermal management strategies available to the U.S. repository program, given the range of disposal concepts. A host medium such as salt with greater thermal conductivity and peak temperature tolerance could shorten decay storage by 50 years, or facilitate the use of larger waste packages. The LWR UOX SNF evaluated in this study represents that which could be produced in the coming decades. The existing, lower burnup used fuel that is presently in storage at many LWR locations across the U.S. is significantly cooler, and analyses of this type could be used to show that disposal is possible with less decay storage or larger waste packages. We note that while the temperature limits and waste package capacities used in this study are similar to those used internationally and in past U.S. studies, they might be increased as the result of ongoing research and development activities. This study selected enclosed emplacement modes to conform with disposal concepts developed internationally and previously in the U.S. Open modes (such as that proposed for a repository at Yucca Mountain) afford additional flexibility in waste management and the necessary investment, because the same facility serves both storage and disposal functions. Use of open modes, and combined analysis of storage, transportation, and disposal functions, are appropriate to consider in future studies of this type. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

ARCADIA and advanced methods licensing and implementation update

AREVA NP is implementing a consistent code and methodology suite for core design and safety analysis, based on first principles modeling with an extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include capabilities that were not available in the past. An example is the 3D coupled transient core power distribution characterization, which allows more realistic modeling of plant behavior, removal of inaccuracies by high resolution and access to local information. Another one is the progressive replacement, respectively complement of old deterministic methods by order statistics. This paper provides an update of the current licensing and implementation status of the main AREVA NP new codes (ARCADIA, GALILEO, S-RELAP5) and their associated advanced methods for PWR applications. ARCADIA is the state-of-the-art neutron transport/ thermal-hydraulics (TH)/ thermal-mechanics (TM) code system for PWR applications. Development is completed, implementation started in France, the Netherlands, the U.S.A. and other countries. The extension to BWR applications is on-going. GALILEO is the state-of-the-art fuel rod performance code for PWR and BWR applications. Development is completed, implementation started in France and the U.S.A. S-RELAP5 is a thermal-hydraulic code that has been completely re-coded and modernized in terms of software architecture. S-RELAP5 has been successfully coupled with ARCADIA. It is shown that advanced codes and integrated methods for accident analysis are keys for extending the fuel cycle design space.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Position-Dependency of Fuel Assembly Homogenization in a Pressurized Water Reactor

2-step nuclear reactor core analysis has long been established as the conventional calculational procedure for the pressurized water reactor (PWR) due to its efficiency. This is because the procedure significantly avoids the potentially huge computational burden arising from an explicit geometrical modeling and consideration of over ∼ 500 energy groups. It instead approaches the problem with spatial homogenization and group condensation: a 2-D lattice transport calculation to homogenize the fuel assembly cross-sections and a 3-D whole core nodal diffusion. One notes that the few-group constant generation using deterministic method could be complicated due to its limited flexibilities in geometrical modeling and inherent drawbacks of the multi-group approximation. Furthermore, accuracy of the deterministic few-group constant generation can especially be compromised during the resonance treatment calculation. As such, an alternative lattice calculation using the Monte Carlo (MC) approach, in place of the deterministic fuel assembly homogenization (few-group constant generation), has previously been proposed. During the assembly homogenization, directions and magnitudes of the current at boundaries are unknowns. Therefore, zero net current boundary conditions (infinite lattice) are instead used as the assembly node. The assumption is quite reasonable considering heterogeneities inside the node are more dominant than heterogeneities due to nodal location in the core. In fact, this approximation has successfully been applied to many LWR core analyses to reasonable accuracy. It should nonetheless be noted that the net current zero boundary assumption effectively ignores position-dependency of the fuel assembly homogenization in the general 2-step method. As such, there are a number of well-known researches on the extended assembly homogenization techniques, which introduce correction factors to compensate effects of the assembly environment, such as Generalized Equivalence Theory (GET), Super-homogenization method (SPH), and color set calculation method. Even though the assembly environment effect has long been acknowledged, its effect has not properly been investigated due to the limited computing power in the past and due to the reasonable accuracy of the extended assembly homogenization techniques. Nonetheless, quantification of position-dependency of the assembly homogenization in the core could still potentially offer additional value in terms of calculation accuracy to the 2-step reactor analysis. It is upon this motivation that the paper is prepared. In this research, position-dependent few-group constants were calculated and compared against those of infinite lattices. All calculations were completed using the MC Serpent2 code with ENDF/B7.0 nuclear data library. In this study, position-dependency of the fuel assembly homogenized cross sections in a PWR core has been investigated through comparison of the conventional 2-step reactor analysis with the reference high fidelity Monte Carlo simulation. It was found that the homogenized two-group constants are clearly position-dependent as the discrepancies between infinite lattice and whole-core calculations are quite significant especially in a rodded assembly. As such, the corresponding correction required due to the assembly environment effect is not negligible compared to the correction required due to assembly heterogeneity itself. It is therefore quite clear that position-dependency of the group cross sections should be considered in the 2-step procedure for a more accurate PWR core analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Power peaking with small modular reactors to support deep penetration of renewable energy sources - 192

As inherently intermittent sources of renewable energy more fully penetrate the US grid, peaking power is largely being supplied by carbon-emitting natural gas turbines. Significant greenhouse gas emissions can be avoided if these plants are replaced with carbon-neutral nuclear facilities to provide peaking power to complement renewable generation and meet overall power demand. There is a great deal of previous work regarding reactor power shaping with control rod movement for both currently operating nuclear power plants and proposed plants, but the literature on load-following to meet less predictable, more rapidly varying power demand is less comprehensive. We have chosen the Westinghouse IRIS integral PWR as our candidate for modeling, simulation, and control studies. A nodal model of the IRIS integral PWR, partially adapted from a MATLAB/Simulink model constructed by prior students at the University of Tennessee in Knoxville, has been constructed in Modelica. Currently the model includes the reactor vessel and steam generator system and simple assumptions for the balance of plant. The control scheme for the load-following operation of an IRIS integral PWR model would ultimately lead to the development of real operational mechanisms and principles in a grid with significant renewables capacity. In the future the model will be integrated with sophisticated grid and hybrid energy systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

CTF Parallel Performance Improvements

CTF is a modernized version of COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid), which is a thermal hydraulic (T/H) sub-channel code based on a two-phase, two-fluid model that formulates the conservation equations of mass, energy, and momentum for three fields of vapor, continuous liquid, and entrained liquid droplets. Under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program sponsored by the U.S. Department of Energy, CTF is being further developed and improved for modeling and simulation of light water reactors. The code improvements for predicting T/H responses of pressurized water reactor (PWR) cores are focused on its applications to specific challenge problems deemed significant to the industry, such as departure from nucleate boiling (DNB) and crud-induced power shift. The code is also being used for T/H feedback in normal operating condition depletions in PWR and boiling water reactor designs in the reactor core simulator, Virtual Environment for Reactor Applications Core Simulator (VERA-CS), being developed by CASL. CTF modifications under the CASL program include software optimization, new closure models, and parallelization for modeling full reactor core T/H responses. The purpose of this work was to improve the parallel performance of CTF. CTF was originally parallelized via a domain-decomposition strategy using the Message Passing Interface (MPI), in which the basis of a solution domain was a physical assembly in the model. In other words, a user who wanted to perform a parallel solution with CTF had to provide the code one processor per assembly in the model (e.g., 56 cores for a 56 assembly quarter-core model, or 193 cores for a 193 assembly full-core model). This design was rigid, demanding a very specific number of processors for a model; furthermore, when CTF was run in VERA-CS-which typically uses thousands of processors- many processors sat idle during the CTF solve because CTF was unable to use any more processors than there are physical assemblies. This was not necessarily a problem if the CTF execution speed was very fast, but the average time spent in a CTF solve for prototypical PWR operating conditions was significant (i.e., over 20 % of the simulation wall time was being spent in CTF for a standard core depletion simulation). Larger, more computationally intensive simulations performed in support of the CASL DNB challenge problem demonstrated a need for improvements in this original parallelization approach. The specific goal of this work was to allow CTF to use more processors than there are physical assemblies in the model, which would not only speed up CTF and improve the overall efficiency of VERA-CS but also give users more-flexible options for executing their simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Steam Accumulator Storage Integration into a Nuclear Power Plant

The growth of non-dispatchable renewables benefits technologies with flexible load capabilities. Baseload technologies must accept low prices and revenues during low-demand periods that arise with increasing frequency as renewables deployment grows. For this reason, generators such as combined cycle gas turbines threaten to out-compete baseload, and nuclear power plants are facing early retirement. If utilities are to consider future nuclear power plants to be attractive investments they must excel at roles other than baseload generation. Energy storage offers a means to avoid times of low prices and capitalize on times of high prices, as well as for nuclear to participate in ancillary markets and provide grid stability and regulation services. This paper explores the use of steam accumulators to store energy for a pressurized water reactor (PWR). The discharging steam from the accumulator provides heat to a feedwater heater, replacing steam, which bleeds off of the high-pressure turbine during normal operation. The utilization of the accumulator as a heat source for the feedwater heater increases the electrical power output of the PWR during discharge, as more electricity can be produced for a constant thermal input from the steam generator. This study first presents the proposed layout of the steam accumulator integrated into a PWR and the base case values used. Next a brief overview of the thermodynamic model governing the charging and discharging processes of the accumulator is presented. An economic model describes the capital cost estimation and revenue potential for the system. These models are evaluated at varying values of energy, power, and pipe lengths. Results show the relationship between energy, power, and accumulator size, and their effect on net annual revenue and storage efficiency. The cases presented here provide direction for future work for optimizing the size and power output of the accumulator.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Use of tsunami in the validation of scale 6.1 for the Swedish spent fuel repository - 14381

In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent Fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several updates to the application. One of SSM's concerns was the validation of the codes. SSM emphasized that SKB must improve the method to select critical experiments to the validation suite. To satisfy this demand, SKB used the tool Tsunami in the Scale 6.1 code package to select experiments from the handbook of criticality experiment (IHECSBE). With help of sensitivity files SKB evaluated almost 700 experiments from the LEU-COMP-THERM library. SKB then picked the experiments that most resembled our safety cases. The sensitivity coefficient, ck, from Tsunami has been used to evaluate the similarity between the experiments and the applications. It has been considered that a ck over 0.8 provides a 'good enough' similarity between the experiment and our evaluated safety cases. For the different safety cases between 50-600 experiments satisfied the 0.8 criteria giving a good base of data for statistical analysis. It is worth noticing that PWR-fuel generally receives a higher ck than BWR fuel. Initial analysis shows that the differences in results between BWR and PWR can be explained by a larger sensitivity for the U{sup 238} n,n' part of the cross section in the PWR cases. To make sure the validation covers all materials and required physical parameters a gap analysis was made using engineering judgement by criticality safety experts. From this exercise additional experiments were added to the suite. This paper will discuss the Tsunami results for different safety cases and the reasons for differences between them. It will also present the final results of the validation and discuss experiments that were excluded from the validation suite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗